Draft 

ADVISORY BOARD ON 

RADIATION AND WORKER HEALTH 

National Institute for Occupational Safety and Health 

ARGONNE NATIONAL LABORATORY-WEST SEC-00224

REACTOR PRIORITIZATION FOR EVALUATION OF 

ORAUT-OTIB-0054 APPLICABILITY

Contract No. 211-2014-58081 

SCA-TR-2016-SEC010, Revision 0 

Prepared by 

Stephen L. Ostrow, PhD

SC&A, Inc.

1608 Spring Hill Road, Suite 400 

Vienna, Virginia 22182 

Saliant, Inc.

5579 Catholic Church Road 

Jefferson, Maryland 21755 

July 2016 

DISCLAIMER

This is a working document provided by the Centers for Disease Control and Prevention (CDC) technical 

support contractor, SC&A for use in discussions with the National Institute for Occupational Safety and 

Health (NIOSH) and the Advisory Board on Radiation and Worker Health (ABRWH), including its 

Working Groups or Subcommittees. Documents produced by SC&A, such as memorandum, white paper, 

draft or working documents are not final NIOSH or ABRWH products or positions, unless specifically 

marked as such. This document prepared by SC&A represents its preliminary evaluation on technical 

issues.

NOTICE: This report has been reviewed to identify and redact any information that is protected by the 

Privacy Act 5 U.S.C. § 552a and has been cleared for distribution.
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SC&A, INC.: Technical Support for the Advisory Board on Radiation & 

Worker Health Review of NIOSH Dose Reconstruction Program

DOCUMENT TITLE: 

Argonne National Laboratory-West SEC-00224 Reactor Prioritization 

for Evaluation of ORAUT-OTIB-0054 Applicability

DOCUMENT NUMBER/

DESCRIPTION:

SCA-TR-2016-SEC010 

REVISION NO.: 0 (Draft)

SUPERSEDES: N/A

EFFECTIVE DATE: July 13, 2016

TASK MANAGER: John Stiver, MS, CHP [signature on file]

PROJECT MANAGER: John Stiver, MS, CHP [signature on file]

DOCUMENT 

REVIEWER(S):

John Stiver, MS, CHP [signature on file]

John Mauro, PhD, CHP [signature on file]

Record of Revisions

Revision 

Number

Effective 

Date Description of Revision

0 (Draft) 07/13/2016 Initial issue
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TABLE OF CONTENTS

Abbreviations and Acronyms ......................................................................................................... 4

1.0 Introduction.......................................................................................................................... 7

2.0 ORAUT-OTIB-0054 ............................................................................................................ 9

3.0 Evaluation .......................................................................................................................... 11

4.0 References .......................................................................................................................... 14

Attachment 1. ANL-W Reactor Prioritization with Respect to ORAUT-OTIB-0054 Applicability 

Investigation.................................................................................................................................. 16
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ABBREVIATIONS AND ACRONYMS

Advisory Board Advisory Board on Radiation and Worker Health 

AFSR Argonne Fast Source Reactor 

ALPR Argonne Low Power Reactor 

ANL-W Argonne National Laboratory-West

ARA Auxiliary Reactor Area

ARMF Advanced Reactivity Measurement Facility Number

ATR Advanced Test Reactor

ATRC Advanced Test Reactor Critical Facility

BORAX Boiling Water Reactor Experiment

BWR boiling water reactor 

CET Critical Experiment Tank

CFA Central Facilities Area

CFRMF Coupled Fast Reactivity Measurement Facility

CRCE Cavity Reactor Critical Experiment

Cs cesium

D diameter

DOE U.S. Department of Energy 

EBOR Experimental Beryllium Oxide Reactor

EBR Experimental Breeder Reactor

EOCR Experimental Organic Cooled Reactor

ETR Engineering Test Reactor

ETRC Engineering Test Reactor Critical Facility

F Fahrenheit

FFTF Fast Flux Test Facility

FRAN Nuclear Effects Reactor 

GCRE Gas Cooled Reactor Experiment

GWth gigawatt thermal

H height

HFEF Hot Fuel Examination Facility

HOTCE Hot Critical Experiment
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HTRE Heat Transfer Reactor Experiment

IETF Initial Engine Test Facility

INL Idaho National Laboratory 

kWe kilowatt electric

kWth kilowatt thermal

LMFBR Liquid Metal Fast Breeder Reactor

LOFT Loss of Fluid Test Facility

LPTF Low Power Test Facility

ML-1 Mobile Low-Power Reactor-1 

MTHM metric ton heavy metal

MTR Materials Test Reactor

MTU metric ton uranium 

MW megawatt

MWd megawatt days

MWe megawatt electric

MWth megawatt thermal

Na sodium 

NaK sodium-potassium (liquid metal)

NRAD Neutron Radiography Facility 

NRF Naval Research Facility

NIOSH National Institute for Occupational Safety and Health

OMRE Organic Moderated Reactor Experiment

ORAU(T) Oak Ridge Associated Universities (Team) 

ORNL Oak Ridge National Laboratory 

OTIB ORAUT Technical Information Bulletin

PBF Power Burst Facility

psig pounds per square inch gauge 

Pu plutonium 

RMF Reactivity Measurement Facility

SCRCE Spherical Cavity Reactor Critical Experiment

SEC Special Exposure Cohort 
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SEFOR Southwest Experimental Fast Oxide Reactor

SL-1 Stationary Low-Power Reactor

SNAP Systems for Nuclear Auxiliary Power

SNAPTRAN Systems for Nuclear Auxiliary Power Transient

SPERT Special Power Excursion Reactor Test

Sr strontium 

STPF Shield Test Pool Facility

STR Submarine Thermal Reactor

SUSIE Shield Pool Test Facility Reactor

TAN Test Area North

THRITS Thermal Reactor Idaho Test Station

TRA Test Reactor Area

TREAT Transient Reactor Test Facility

TRIGA Training, Research, Isotope General Atomics (reactor)

U uranium 

UO2 uranium dioxide 

WRRTF Water Reactor Research Test Facility

ZPPR Zero Power Physics Reactor

ZPR Zero Power Reactor
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1.0 INTRODUCTION

One of the topics presented at the July 23–24, 2014, meeting of the Advisory Board on Radiation 

and Worker Health (hereafter referred to as the “Advisory Board”) in Tampa, Florida, was the 

Special Exposure Cohort (SEC) Petition SEC-00224 evaluation report for Argonne National 

Laboratory-West (ANL-W) (NIOSH 2016) prepared by the National Institute for Occupational 

Safety and Health (NIOSH) and its technical contractor, Oak Ridge Associated Universities 

(ORAU). At that meeting, the Advisory Board charged SC&A with reviewing the evaluation 

report.

Part of the review concerns the issues of dose reconstructability and gap analysis. With respect to 

the latter, inherent in the SEC framework is the assumption that doses can be reconstructed with 

sufficient accuracy for site areas and time periods that lie outside the SEC class definition and 

that are not being held in reserve for further evaluation by NIOSH. Operations at ANL-W 

involving radioactive materials were very complex, as many unique nuclear reactors and 

experiments were built and tested.

A primary tool that NIOSH uses for internal dose reconstruction is the guidance appearing in 

ORAUT-OTIB-0054, Fission and Activation Product Assignment for Internal Dose-Related 

Gross Beta and Gross Gamma Analyses (hereafter referred to as “OTIB-0054”) (ORAUT 2015). 

Except for certain situations, OTIB-0054 assigns fission and activation product intakes for 

different radioisotopes that are directly tied to an indicator radionuclide (strontium-90 [Sr-90] or 

cesium-137 [Cs-137]). OTIB-0054 generated nine different representative reactor cases, which 

are intended to envelope the range of reactor and nuclear fuel types and operating scenarios to 

which workers might have been exposed. 

Before examining each ANL-W reactor in detail and determining whether its operations were 

bounded by the methodology in OTIB-0054, SC&A prioritizes in this report the reactors for 

further investigation, where one of the main factors is the degree to which the abundance of 

fission and activation products and actinides relative to the abundance of Cs-137 and Sr-90 bear 

any resemblance to the mix of radionuclides in OTIB-0054.1

1 A similar process was followed for the INL reactors, where INL SEC-00219 Reactor Prioritization for 

Evaluation of ORAUT-OTIB-0054 Applicability (SC&A 2016) prioritized them for further examination, NIOSH 

SEC-00219: Test Reactor Area Modeling (SC&A 2015a) examined the dose reconstructability assumption for the 

major reactors in the Test Reactor Area, and Review of NIOSH Strategy for Reconstructing Internal Doses to 

Workers at Test Area North (SC&A 2015b) examined dose reconstructability for some of the reactors in Test Area 

North. 

Susan Stacy, in her comprehensive review of the history of the Idaho National Laboratory (INL) 

from inception through 1999, Proving the Principle: A History of the Idaho National 

Engineering and Environmental Laboratory (Stacy 2000), lists in Appendix B the 52 reactors 

that were built on the INL site (including two that never operated) and provides a brief summary 

of each. SC&A, following the practice of NIOSH in its INL/ANL-W reports (e.g., the site 

profile), uses Stacy’s list as a convenient framework to examine the various reactors.

Attachment 1 lists all the reactors and, for each covered by the ANL-W SEC petition, notes its 

 
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operating period, provides a brief summary description, and presents SC&A’s screening 

assessment of its priority ranking with respect to performing a detailed review on whether 

OTIB-0054 bounds its makeup and operating conditions. 

The priority rankings are divided into three categories: High, Medium, and Low. Though based 

on a substantial amount of research, the rankings are still somewhat subjective because a full 

analysis would involve detailed and extensive research for each reactor and actually performing 

the OTIB-0054 applicability analyses themselves, which would go counter to the limited 

objectives of this screening process. The assignment of reactors to priority ranking categories 

considers reactor design factors such as the type of fuel (e.g., solid or gaseous, uranium or 

plutonium-based), enrichment (e.g., low-enriched commercial-type fuel or fully-enriched fuel), 

cladding (e.g., aluminum or steel), moderator (e.g., light water, heavy water, or beryllium), 

coolant (e.g., water or liquid metal), operational mode (e.g., steady-state or periodic), length of 

operation, and whether the reactor performed within design limits or was deliberately or 

inadvertently taken outside those limits (e.g., in tests supporting power reactor safety programs). 

Also considered qualitatively is the potential for significant radiation exposure of personnel. 

These screening criteria were selected because they were judged by SC&A to be those criteria 

that would be best indicative of the degree to which the default mix of radionuclides in 

OTIB-0054 might result in an underestimate of the internal doses to workers, or simply result in 

unrealistic estimates of the internal doses to workers at ANL-W who worked in the vicinity of 

these reactors or worked with irradiated fuel from these reactors.

In addition to OTIB-0054 applicability, this report also considers the following four factors, to 

the extent that they are known, that reflect the scope of the population potentially “at risk” of 

uncontrolled/unmonitored exposures. 

• Duration reactor was in operation 

• Frequency/intensity of operation 

• Where possible, the approximate number of workers potentially exposed during its 

operation 

• Incidents or other factors with potential to contribute to the risk of 

unintended/unprotected exposures

Of the 52 reactors in the entire INL site, 12 reactors operated at ANL-W; these are listed in 

Table 3. 
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2.0 ORAUT-OTIB-0054

After a series of initial runs using the ORIGEN2 isotope generation and depletion code 

(Croff 1980), OTIB-0054 selects four actual reactors to represent different general categories of 

reactors that might envelope the wide variety of reactors at the different sites considered in the 

Energy Employees Occupational Illness Compensation Program Act of 2000 program.2

2

 ORAUT 2015 should be consulted for the details of the selection process. 

The 

representative reactors are listed in Table 1.

Table 1. ORAUT-OTIB-0054 Representative Reactors

Category Reactor

High-flux reactors Advanced Test Reactor (ATR)

Na-cooled fast reactors Fast Flux Test Facility (FFTF)

Pu production reactors Hanford N-Reactor

Research reactors TRIGA with stainless steel cladding

Source: ORAUT 2015.

Multiple ORIGEN-S (ORNL 2015) runs performed by NIOSH produced a total of nine 

representative cases for the four reactors. ORIGEN-S is a more modern and capable version of 

ORIGEN and is part of the SCALE code system (ORNL 2015) for nuclear safety analysis and 

design, developed and maintained by the Oak Ridge National Laboratory (ORNL) for the 

U.S. Nuclear Regulatory Commission. Table 2 (OTIB-0054 Table 5-2) lists the parameters and 

basis selected by NIOSH for each of the cases.

 
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Table 2. ORIGEN-S Irradiation Parameters for the Nine Representative Reactor Cases

Case Parameters Basis

ATR 1 

Specific power = 2,379.1 MW/MTU

Irradiation time = 132.27 days

Burnup = 314,684 MWd/MTU

Maximum burnup at nominal power. 

ATR 2 

Specific power = 8,651.2 MW/MTU

Irradiation time = 36.4 days

Burnup = 314,904 MWd/MTU

Maximum burnup at maximum assembly

power. 

ATR 3 

Specific power = 2,379.1 MW/MTU

Irradiation time = 56 days 

Burnup = 133,230 MWd/MTU

Nominal burnup at nominal power. 

FFTF 1 

Specific power = 163.8

MW/MTHM Irradiation time = 

929.4 days

Maximum burnup at nominal power. 

FFTF 2 

Specific power = 163.8

MW/MTHM Irradiation time = 

488.3 days

Nominal burnup at nominal power. 

N Reactor 1 

Specific power = 10.4 MW/MTU

Irradiation time = 114.2 days 

Burnup = 1,188 MWd/MTU

Production of weapons-grade plutonium

(nominal 6% Pu-240 content) at nominal 

power.

N Reactor 2 

Specific power = 10.4 MW/MTU

Irradiation time = 285.6 days 

Burnup = 2,970 MWd/MTU

Production of fuel-grade plutonium

(nominal 12% Pu-240 content) at nominal 

power.

TRIGA 1 

Specific power = 15.57 MWd/MTU

Irradiation time = 730.1 days 

Burnup = 11,368 MWd/MTU

Maximum burnup at nominal power. 

TRIGA 2 

Specific power = 15.57 MW/MTU

Irradiation time = 115.2 days 

Burnup = 1994 MWd/MTU

Nominal burnup at nominal power. 

Source: Reproduced from ORAUT 2015, Table 5-2.
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3.0 EVALUATION

NIOSH uses OTIB-0054 to determine internal doses to claimants using indicator radionuclides in 

cases where only gross beta or gross gamma measurements are available. The nine cases of 

Table 2 are intended to envelope reactor and nuclear fuel types and operating scenarios to which 

workers might have been exposed. As discussed in Section 1, the screening process in this report 

examines whether it seems likely that OTIB-0054 adequately envelopes the ANL-W reactors 

listed in Attachment 1 and prioritizes for further investigation those that might not be enveloped 

by the OTIB-0054 methodology. Table 3 lists all 52 reactors alphabetically (which is also 

numerically), as shown in Stacy (2000). The non-ANL-W reactors are included for convenience 

in identifying a particular reactor in the overall INL site. The rows for the non-ANL-W reactors 

are shaded. Priority rankings for the ANL-W reactors from Attachment 1 are also included in the 

last column.

Table 3. List of all INL Reactors 

Reactor Name a,b,c Location Priority 

Ranking

1. Advanced Reactivity Measurement Facility No. 1 

(ARMF-I)

TRA (Bldg. TRA-660) —

2. Advanced Reactivity Measurement Facility No. 2 

(ARMF-II). Renamed Coupled Fast Reactivity Measurement 

Facility (CFRMF) in 1968

TRA (Bldg. TRA-660) —

3. Advanced Test Reactor (ATR) TRA (Bldg. TRA-670) —

4. Advanced Test Reactor Critical Facility (ATRC) TRA (Bldg. TRA-670) —

5. Argonne Fast Source Reactor (AFSR) ANL-W Low

6. Boiling Water Reactor Experiment No. 1 (BORAX-I) ANL-W High

7. Boiling Water Reactor Experiment No. 2 (BORAX-II) ANL-W High

8. Boiling Water Reactor Experiment No. 3 (BORAX-III) ANL-W High

9. Boiling Water Reactor Experiment No. 4 (BORAX-IV) ANL-W High

10. Boiling Water Reactor Experiment No. 5 (BORAX-V) ANL-W High

11. Cavity Reactor Critical Experiment (CRCE) TAN - WRRTF – LPTF —

12. Coupled Fast Reactivity Measurement Facility 

(CFRMF). Formerly named Advanced Reactivity 

Measurement Facility No. 2 (ARMF-II)

TRA (Bldg. TRA-660) —

13. Critical Experiment Tank (CET) TAN - WRRTF – LPTF —

14. Engineering Test Reactor (ETR) TRA (Bldg. TRA-642) —

15. Engineering Test Reactor Critical Facility (ETRC) TRA (Bldg. TRA-654) —

16. Experimental Beryllium Oxide Reactor (EBOR) TAN - WRRTF – LPTF 

(Bldg. TAN-646)

—

17. Experimental Breeder Reactor No. I (EBR-I) ANL-W High

18. Experimental Breeder Reactor No. II (EBR-II) ANL-W High

19. Experimental Organic Cooled Reactor (EOCR) CFA (vicinity) —

20. Fast Spectrum Refractory Metals Reactor (710) TAN - WRRTF – LPTF —

21. Gas Cooled Reactor Experiment (GCRE) ARA-III —

22. Heat Transfer Experiment No. 1 (HTRE-1) TAN - IETF —

23. Heat Transfer Experiment No. 2 (HTRE-2) TAN - IETF —

24. Heat Transfer Experiment No. 3 (HTRE-3) TAN - IETF —

25. High Temperature Marine Propulsion Reactor (630-A) TAN - WRRTF – LPTF —

26. Hot Critical Experiment (HOTCE) TAN - WRRTF – LPTF —
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Reactor Name a,b,c Location Priority 

Ranking

27. Large Ship Reactor A (A1W-A) NRF —

28. Large Ship Reactor B A1W-B NRF —

29. Loss of Fluid Test Facility (LOFT) TAN (Bldg. TAN-650) —

30. Materials Test Reactor (MTR) TRA (Bldg. TRA-603) —

31. Mobile Low-Power Reactor No. 1 (ML-1) ARA-IV —

32. Natural Circulation Reactor (S5G) NRF —

33. Neutron Radiography Facility (NRAD) ANL-W Low

34. Nuclear Effects Reactor (FRAN) ARA-IV —

35. Organic Moderated Reactor Experiment (OMRE) East of CFA, between Waste Area 

Groups 4 and 5

—

36. Power Burst Facility (PBF) Near the SPERT-I site. —

37. Reactivity Measurement Facility (RMF) TRA (Bldg. TRA-603) —

38. Shield Test Pool Facility (STPF - SUSIE) TAN - WRRTF – LPTF 

(Bldg. TAN-646)

—

39. Special Power Excursion Reactor Test No. I (SPERT-I) Separate complex east of CFA —

40. Special Power Excursion Reactor Test No. II (SPERT-II) Separate complex east of CFA —

41. Special Power Excursion Reactor Test No. III (SPERTIII)Separate complex east of CFA —

42. Special Power Excursion Reactor Test No. IV (SPERTIV)Separate complex east of CFA —

43. Spherical Cavity Reactor Critical Experiment (SCRCE) TAN – WRRTF - LPTF —

44. Stationary Low-Power Reactor (Earlier name: Argonne 

Low Power Reactor) (SL-1, ALPR) 

ARA-II —

45. Submarine Thermal Reactor (S1W, STR). Also known 

as the Submarine Prototype Reactor.

NRF —

46. Systems for Nuclear Auxiliary Power (SNAP) 10A 

Transient No. 1 (SNAPTRAN–1)

TAN - IETF —

47. Systems for Nuclear Auxiliary Power (SNAP) 10A 

Transient No. 3 (SNAPTRAN–3)

TAN - IETF —

48. Systems for Nuclear Auxiliary Power (SNAP) 10A 

Transient No. 2 (SNAPTRAN–2)

TAN – IETF —

49. Thermal Reactor Idaho Test Station (THRITS) TAN – WRRTF - LPTF —

50. Transient Reactor Test Facility (TREAT) ANL-W Medium

51. Zero Power Physics Reactor (Earlier name: Zero Power 

Plutonium Reactor) ZPPR 

ANL-W Low

52. Zero Power Reactor No. 3 (ZPR-III) ANL-W Low

Notes: 

(a) The list of 52 reactors was taken from Stacy 2000. 

(b) Location acronyms (current names are used in most cases): ANL-W = Argonne National Laboratory-West; 

ARA = Auxiliary Reactor Area; CFA = Central Facilities Area; IETF = Initial Engine Test Facility; LPTF = Low 

Power Test Facility; NRF = Naval Reactor Facility; TAN = Test Area North; TRA = Test Reactor Area; 

WRRTF = Water Reactor Research Test Facility. 

(c) Shaded rows indicate non-ANL-W reactors at the INL site. 

Attachment 1 categorizes the 12 ANL-W reactors according to three prioritization levels: High, 

Medium, or Low, as discussed in Section 1. It is apparent from looking at the summary 

descriptions of Attachment 1 that all of the ANL-W reactors were different from the four 

representative reactors of OTIB-0054 to a significant degree, since most of the former were oneof-a-kind experiments that might have utilized different fuel types (e.g., fissile materials, 
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chemical forms, cladding, and physical arrangements), blankets (e.g., to breed more fuel), 

moderators, and coolants, and were often deliberately or inadvertently operated beyond design 

limits, sometimes to failure. In addition, rather than operate at more-or-less steady-state 

conditions for some length of time, some ANL-W reactors operated in pulsed mode, in which 

they produced a huge amount of power in a very short time interval before they shut themselves 

down (due to strongly negative reactivity properties, such as negative void, temperature, or 

expansion coefficients), or in intermittent mode, where a series of experiments were run, then the 

reactor shut down and possibly modified until the next series of experiments. Fuel burnups were 

frequently considerably lower than for the representative reactors of OTIB-0054, so that the 

long-lived decay products did not have the chance to build up in the fuel, resulting in different 

isotopic ratios than in the OTIB-0054 reactors. In addition, many of the individual ANL-W 

reactors experimented with different fuels, core configurations, coolants, moderators, etc. over 

their operating lifetimes.

Notwithstanding the above considerations, SC&A categorized the 12 ANL-W reactors as a guide 

for determining which reactors should be considered first in a more detailed study, such as was 

done in SC&A 2015a for some reactors in the Test Reactor Area (TRA) and in SC&A 2015b for 

some in Test Area North (TAN). The results of the categorizations of Attachment 1 are 

summarized in Table 4, which shows that seven reactors were put in the High category, one in 

the Medium category, and four in the Low category. High rankings were assigned in cases where 

it was thought that the mix of beta emitters from a particular reactor’s fuel would be very 

different than the mix in the OTIB-0054 reactors, such that doses might be materially 

underestimated. This would not be evident, however, until the products of the relative amount of 

each radionuclide times its inhalation dose conversion factor, times its release fraction are 

calculated and added up. This index of risk would be compared to that associated with the mix in 

OTIB-0054. Only through such a process would it be determined if OTIB-0054 is bounding, but 

that determination would only be made in a detailed examination, which would be the next step 

beyond this report’s screening assessment, which must rely on heuristic arguments.

The rationale behind the categorizations shown in Table 4 appears in Attachment 1. SC&A 

recommends the natural groupings of the five Boiling Water Reactor Experiment (BORAX) 

reactors together and the two Experimental Breeder Reactors (EBRs) together, all in the High 

group. 

Table 4. Priority Class Categorization for ANL-W 

Priority Class Reactor Number from Attachment 1

High (7) 6, 7, 8, 9, 10, 17, 18

Medium (1) 50

Low (4) 5, 33, 51, 52

Total Reactors 12
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4.0 REFERENCES

Amundson 1969. P. Amundson, W. Davey, R. Currah, G. Brunsop, and R. Huber, Reactor 

Physics Measurements in the Argonne Fast Source Reactor, Argonne National Laboratory, Idaho 

Falls, Idaho. 1969. Available at http://www.osti.gov/scitech/servlets/purl/4764378. 

ANL 2016. Reactors Designed by Argonne National Laboratory. Last modified October 2, 2014. 

Available at http://www.ne.anl.gov/About/ANL-Reactors.shtml. 

Brunson 1959. G. Brunson, Design and Hazards Report for the Argonne Fast Source Reactor 

(AFSR), ANL-6024. June 1959. Available at http://www.osti.gov/scitech/servlets/purl/4203554. 

Bumgardner 2014. J. Bumgardner, Restart of the Transient Reactor Test Facility (TREAT) and 

Resumption of Transient Testing, Idaho National Laboratory. November 2014. Available at 

http://www.cab.cnea.gov.ar/igorr2014/images/presentations/18thNovTuesday/CondorRoom/1stB

lock/01JohnBUMGARDNER.pdf. 

Craft et al. 2015. A. Craft, D.M. Wachs, M.A. Okuniewski, D.L. Chichester, W.J. Williams, 

G.C. Papaioannou, and A.T. Smolinski, “Neutron Radiography of Irradiated Nuclear Fuel at 

Idaho National Laboratory, 10th World Conference on Neutron Radiography, 5-10 October 

2014,” Physics Procedia 69 (2015): 483–490. Available at 

http://www.sciencedirect.com/science/article/pii/S1875389215006781. 

Croff 1980. A.G. Croff, ORIGEN2 – A Revised and Updated Version of the Oak Ridge Isotope 

Generation and Depletion Code, ORNL-5621, Oak Ridge National Laboratory. July 1980. 

[SRDB Ref. ID 23569] 

Davey 1964. W. Davey, “An Analysis of 23 ZPR-III Fast Reactor Critical Experiments,”

Nuclear Science and Engineering, 19 (3): 259–273. July 1964. 

INL 2016a. Experimental Breeder Reactor-I. INL Fact Sheet, Idaho National Laboratory, 

accessed June 29, 2016. Available at http://www4vip.inl.gov/research/experimental-breederreactor-1/d/experimental-breeder-reactor-1.pdf. 

INL 2016b. Transient Reactor Test Facility. INL Fact Sheet, Idaho National Laboratory, 

accessed June 29, 2016. Available at http://www4vip.inl.gov/research/transient-reactor-testfacility/d/transient-reactor-test-facility.pdf. 

Lawroski 1972. H. Lawroski, R. Palmer, F. Thalgott, R. Curran, and R. Matlock, Final Safety 

Analysis Report on the Zero Power Plutonium Reactor (ZPPR) Facility, ANL-7471, Applied 

Physics Division, Argonne National Laboratory. June 1972. Available at 

https://www.iaea.org/inis/collection/NCLCollectionStore/_Public/04/056/4056702.pdf. 
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Privacy Act 5 U.S.C. § 552a and has been cleared for distribution.

McFarlane 2010. H. McFarlane, Fast Reactor Operating Experience in the U.S., Idaho National 

Laboratory presentation, March 3, 2010. Available at 

https://www.iaea.org/INPRO/cooperation/4th_GIF_Meeting/32-

McFarlane_US_SFR_Operating_Experience.pdf. 

NIOSH 2016. SEC Petition Evaluation Report: Petition SEC-00224, Revision 0, National 

Institute for Occupational Safety and Health. February 18, 2016.

ORAUT 2010. Idaho National Laboratory and Argonne National Laboratory-West – Site 

Description, ORAUT-TKBS-0007-2, Revision 4, Oak Ridge Associated Universities. August 2, 

2010. 

ORAUT 2015. Fission and Activation Product Assignment for Internal Dose-Related Gross Beta 

and Gross Gamma Analyses, ORAUT-OTIB-0054, Revision 3, Oak Ridge Associated 

Universities. February 6, 2015. 

ORNL 2015. SCALE: A Comprehensive Modeling and Simulation Suite for Nuclear Safety 

Analysis and Design, Oak Ridge National Laboratory. Accessed February 18, 2016, at 

http://scale.ornl.gov. 

SC&A 2015a. NIOSH SEC-00219: Test Reactor Area Modeling, SCA-SEC-2015-0074-C, 

Revision 0, SC&A, Inc., Vienna, Virginia. September 28, 2015. 

SC&A 2015b. Review of NIOSH Strategy for Reconstructing Internal Doses to Workers at Test 

Area North, SCA-TR-2015-SEC0074A, Revision 0, SC&A, Inc., Vienna, Virginia. 

September 28, 2015. 

SC&A 2016. INL SEC-00219 Reactor Prioritization for Evaluation of ORAUT-OTIB-0054 

Applicability, SCA-TR-2016-SEC0002, Revision 1, Vienna, Virginia. June 10, 2016. 

Stacy 2000. Susan M. Stacy, Proving the Principle: A History of the Idaho National Engineering 

and Environmental Laboratory 1949-1999, DOE/ID-10799, Idaho Operations Office, 

U.S. Department of Energy, Idaho Falls, Idaho. 2000. [SRDB Ref. ID 135760]
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ATTACHMENT 1. ANL-W REACTOR PRIORITIZATION WITH RESPECT TO ORAUT-OTIB-0054 

APPLICABILITY INVESTIGATION 

Reactor Name & 

Location a

Operation 

Dates Summary Description b Comments/Priority Ranking 

5. Argonne Fast Source

Reactor (AFSR) 

ANL-W/EBR-I Complex

(1959–1970)/EBR-II 

Complex (1970–late 

1970s)

1959–late 

1970s

The AFSR, a small fast reactor facility, with a design power level of 

1 kilowatt thermal (kWth), fueled with high-enriched uranium, and 

with a blanket of solid depleted uranium with a minimum thickness 

of eight inches, supplied fast and thermal neutrons to: test, calibrate,

and develop radiation detectors for the ANL-W fast reactor program; 

prepare radioactive metallic foils; check out experimental systems 

before placement in other reactors; and develop potential fast reactor 

experiments.

The reactor had a 4.5" diameter (D) × 4.25" height (H) cylindrical 

core of solid high enrichment uranium and a blanket of solid depleted 

uranium (2,100 kg) over eight inches thick. The core was built up 

from uranium discs canned in nickel. Approximately 20 kg of U-235 

were required to achieve criticality.

It had a high density concrete shield and a 4 × 4 × 6 foot graphite 

thermal column. Controls and safety systems were located in a pit 

under the reactor. Experimental beam ports included a graphite 

thermal column and several beam holes.

It was divided into an upper and lower section, with a 1⁄2" D hole at 

the centerline. It occupied two different locations: 1959–1970 – the 

EBR-I Complex in a standalone butler building, and 1970–1997 – the 

EBR-II Complex in a portion of the ZPPR support wing, tied into 

ZPPR ventilation system with negative pressure.

4 to8 people worked periodically at the facility in its first location, 

but significantly more were potentially exposed when it moved to the 

ZPPR facility.

[The text in this cell continues on the next page.]

The AFSR was quite small and lowpowered, so, even though it does not 

appear to be enveloped by the 

representative reactors of OTIB-0054 (it 

operated at a much lower power and 

with a much lower fuel burnup than the 

Fast Flux Test Facility), it doesn’t 

warrant a detailed examination at this 

time. 

Priority Ranking: Low
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Reactor Name & 

Location a

Operation 

Dates Summary Description b Comments/Priority Ranking 

[Row continued from 

previous page; no 

additional text in this 

[Row continued 

from previous 

page; no 

additional text 

in this cell.]

References

ANL 2016

Amundson 1969

Brunson 1959

[Row continued from previous page; no additional 

text in this cell.]

6. Boiling Water

Reactor Experiment No. 

1 (BORAX-I) 

ANL-W/EBR-I Complex

1953–1954 Overview: The BORAX series of reactor experiments, beginning 

with the operation of BORAX-I in 1953, tested the feasibility and 

safety and explored the operating parameters of direct steam 

production in a light water reactor. Would the formation of steam 

bubbles destabilize or stabilize the nuclear reaction? The chronology 

of the BORAX reactors is:

BORAX-I: 7/1953–7/1954 (no surrounding structure)

BORAX-II: 10/1954–3/1955 (new facility)

BORAX-III: 7/1955–4/1956 (modification of B-II)

BORAX-IV: 12/1956–6/1958 (modification of B-III)

BORAX-V: 3/1962–8/1964 (new facility)

BORAX-I: 

The 1.4 megawatt thermal (MWth) reactor was fabricated at Argonne 

National Laboratory-East and shipped to ANL-W, where it was 

assembled and operated. The reactor was water moderated and relied

on natural water cooling and steam formation to remove heat. It was 

housed in a 10-ft diameter and 20-ft deep tank open to the 

atmosphere, which resulted in the requirement of a 1⁄2 mile exclusion 

zone when operating and limited operations to the warmer months of 

the year. BORAX-I was a precursor to the SL-1 plant at INL.

The core was built up from a lower grid and consisted of 26 curved 

plate fuel assemblies of a U-235-aluminum alloy clad with 

aluminum. Ten non-fuel plugs were also present to define the shape 

of the core. [The text in this cell continues on the next page.]

The BORAX series investigated the 

safety of boiling water reactors in five 

different reactors with a variety of 

configurations. The reactors were run in 

both steady-state and in transient modes, 

with the latter leading to core damage in 

some cases. Total burnup for each 

reactor was probably low, since each 

reactor operated for only a short period 

of time and then not continuously. It is 

not apparent whether any of the 

OTIB-0054 reactors could adequately 

envelope the BORAX conditions. It 

appears to make sense to consider the 

five reactors together in one study. 

Priority Ranking: High

cell.]
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Reactor Name & 

Location a

Operation 

Dates Summary Description b Comments/Priority Ranking 

[Row continued from 

previous page; no 

additional text in this cell.]

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continued 

from 

previous 

page; no 

additional 

text in this 

cell.]

Each assembly contained 18 fuel plates joined to aluminum side 

plates to form units approximately 3 inches square and 2 feet long 

(active section). 

Operators initially conducted a series of nondestructive experiments 

consisting of steady-state boiling, as well as over 70 excursion tests 

of <25 ms duration, until the reactor was deliberately destroyed in its 

final experiment on July 22, 1954, through the rapid withdrawal of its 

control rods. Calculations done prior to the test seriously 

underestimated the actual damage, where much of the core melted 

and blew pieces of fuel plates, etc. 200–300 ft away, in a large steam 

explosion and resulting “geyser.”

The BORAX-I testing determined that steam formation was an 

effective and rapid self-limiting feature against short-duration 

excursions due to the large negative coefficient of reactivity. Thus, it 

paved the way for the further development of the boiling water 

reactor (BWR) concept.

References

ANL 2016

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additional text in this cell.]
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Reactor Name & 

Location a

Operation 

Dates Summary Description b Comments/Priority Ranking 

7. Boiling Water

Reactor Experiment No. 

2 (BORAX-II) 

ANL-W/EBR-I Complex

1954–1955 BORAX-II, at 6.4 MWth, continued the boiling water reactor testing 

program beyond BORAX-I. It was contained in a stainless steel 

pressure vessel located below ground in a concrete-lined pit, and

operated with varying enrichments of uranium in its MTR-type fuel 

plates. It was intentionally destroyed in 1955 by taking it prompt 

critical (critical from prompt fission neutrons alone without reliance 

on delayed neutrons, which are released over time from fission 

products). It was much larger than its predecessor and designed to be 

more representative of a future commercial BWR; e.g., it operated at 

a design pressure of 300 pounds per square inch gauge (psig). It 

released excess energy as steam, since it had no turbine generator 

attached. 

The reactor was water cooled (natural circulation), moderated, and 

reflected.

References

ANL 2016

The BORAX series investigated the 

safety of boiling water reactors in five 

different reactors with a variety of 

configurations. The reactors were run in 

both steady-state and in transient modes, 

with the latter leading to core damage in 

some cases. Total burnup for each 

reactor was probably low, since each 

reactor operated for only a short period 

of time and then not continuously. It is 

not apparent whether any of the 

OTIB-0054 reactors could adequately 

envelope the BORAX conditions. It 

appears to make sense to consider the 

five reactors together in one study. 

Priority Ranking: High

8. Boiling Water

Reactor Experiment No. 

3 (BORAX-III) 

ANL-W/EBR-I Complex

1955–1956 The BORAX-II reactor was modified (raising its output to 12 MWth) 

and renamed BORAX-III after the addition of a 2.5 megawatt electric 

(MWe) steam turbine-generator to investigate radioactive 

contamination of the turbine from radioactivity in the primary coolant 

and to demonstrate the production of electricity. The fuel consisted of 

plate-type fuel elements of a 90% enriched uranium-aluminum alloy 

with aluminum cladding.

For two hours on July 17, 1955, the reactor supplied Arco, Idaho with 

500 kilowatt electric (kWe), the BORAX-III facility with 500 kWe, 

and the Central Facilities Area with 1,000 kWe, successfully 

demonstrating power production and distribution from a BWR.

By the time it went out of service in 1956, BORAX-II had operated 

for a total of 1,170 hrs.

References

ANL 2016

The BORAX series investigated the 

safety of boiling water reactors in five 

different reactors with a variety of 

configurations. The reactors were run in 

both steady-state and in transient modes, 

with the latter leading to core damage in 

some cases. Total burnup for each 

reactor was probably low, since each 

reactor operated for only a short period 

of time and then not continuously. It is 

not apparent whether any of the 

OTIB-0054 reactors could adequately 

envelope the BORAX conditions. It 

appears to make sense to consider the 

five reactors together in one study. 

Priority Ranking: High
Effective Date:

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distribution.

Reactor Name & 

Location a

Operation 

Dates Summary Description b Comments/Priority Ranking 

9. Boiling Water

Reactor Experiment No. 

4 (BORAX-IV) 

ANL-W/EBR-I Complex

1956–1958 BORAX-IV, a modification of BORAX-III, at 20 MWth (and 

2.5 MWe) and 300 psig primary coolant pressure, tested uranium

(U-233 and U-235) and thorium ceramic fuel plates (to allow higher 

temperature operations than with uranium fuel plates), some of which 

purposefully contained defects to determine reactor behavior with 

compromised fuel. The tests released some short-lived radionuclides 

to the atmosphere.

Testing at full power with a large number of fuel elements having 

cladding defects, it released approximately 4,565 curies of short-lived 

radionuclides to the atmosphere in March 1958. 

References

ANL 2016

The BORAX series investigated the 

safety of boiling water reactors in five 

different reactors with a variety of 

configurations. The reactors were run in 

both steady-state and in transient modes, 

with the latter leading to core damage in 

some cases. Total burnup for each 

reactor was probably low, since each 

reactor operated for only a short period 

of time and then not continuously. It is 

not apparent whether any of the OTIB0054 reactors could adequately envelope 

the BORAX conditions. It appears to 

make sense to consider the five reactors 

together in one study. 

Priority Ranking: High
Effective Date:

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distribution.

Reactor Name & 

Location a

Operation 

Dates Summary Description b Comments/Priority Ranking 

10. Boiling Water

Reactor Experiment No. 

5 (BORAX-V) 

ANL-W/EBR-I Complex

1962–1964 BORAX-V, a new facility, at 40 MWth, added an integral nuclear 

superheat system to the BORAX experiments to improve thermal 

efficiency of the boiling water reactor plant and to explore the 

feasibility of superheating the steam. The latter raised the 

temperature of saturated steam from 489 °F in the core to 850 °F of 

dry steam and raised the pressure to 600 psig in the superheater.

The reactor system had basically the same configuration as BORAXIV. The reactor was placed in a cylindrical carbon steel pressure 

vessel with ellipsoidal heads that was clad with stainless steel in the 

interior. The reactor was water moderated and water cooled, with the 

superheater section cooled by steam. Cooling could be accomplished 

by natural or forced convection.

The BORAX-V experiments also demonstrated that only negligible 

amounts of radioactivity travelled to the turbine generator section of 

the facility.

References

ANL 2016

The BORAX series investigated the 

safety of boiling water reactors in five 

different reactors with a variety of 

configurations. The reactors were run in 

both steady-state and in transient modes, 

with the latter leading to core damage in 

some cases. Total burnup for each 

reactor was probably low, since each 

reactor operated for only a short period 

of time and then not continuously. It is 

not apparent whether any of the 

OTIB-0054 reactors could adequately 

envelope the BORAX conditions. It 

appears to make sense to consider the 

five reactors together in one study. 

Priority Ranking: High
Effective Date:

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Reactor Name & 

Location a

Operation 

Dates Summary Description b Comments/Priority Ranking 

17. Experimental

Breeder Reactor No. I 

(EBR-I)

ANL-W/EBR-I Complex

1951–1963 The EBR-I fast breeder reactor demonstration was the first reactor 

built at INL and is a Registered National Historic Monument. It had a 

maximum power of 1 MWth from its graphite-reflected, unmoderated,

enriched uranium core. The reactor system was cooled by the liquid 

metal NaK (primary and secondary coolant systems) and surrounded 

by a U-238 breeding blanket. In 1951, it became the first reactor to 

generate electricity from an attached steam turbine-generator, which 

lit a few lightbulbs adjacent to the reactor. The EBR-I also 

demonstrated the feasibility of breeding plutonium in a reactor; i.e., 

creating more nuclear fuel than it consumed in order to supplement 

the scarce supplies of uranium then found in ore bodies (large 

deposits were subsequently found worldwide).

In addition to producing the first electricity from a nuclear plant, 

EBR-I also demonstrated the feasibility of fuel production (breeding), 

whereby fissile Pu-239 was produced in the fertile U-238 blanket. In 

fact, at full power of 1 MWth, 16% of the total power was generated 

in the blanket. Breeding ratios increased from 1.00 in the first core to 

1.27 in the last core. The reactor also had several beam holes for 

instrumentation and irradiation experiments.

The first three of the EBR-I’s four core loadings had highly enriched 

(94% U-235) uranium fuel clad with either stainless steel or 

aluminum. On November 29, 1955, the reactor suffered a 40% to 

50% inadvertent core meltdown due to operator error with control 

rods during an experiment. 

References

INL 2016a 

McFarlane 2016

While OTIB-0054 models the FFTF, a 

fast reactor, and the Hanford N-Reactor, 

a plutonium production reactor, neither 

appears able to adequately envelope the 

EBR reactor conditions of much lower 

power and only intermittent operations. 

Priority Ranking: High
Effective Date:

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Reactor Name & 

Location a

Operation 

Dates Summary Description b Comments/Priority Ranking 

18. Experimental

Breeder Reactor No. II 

(EBR-II) 

ANL-W/EBR-II Complex

1961–1994 The EBR-II, similar to, but a large scale-up from, EBR-I, continued 

fast neutron breeder reactor development at ANL-W, including onsite 

reprocessing of spent fuel into new fuel pins, demonstrating the 

feasibility of a closed fuel cycle. The unmoderated core, with 67% 

enriched U-235 fuel, sat in a tank of 90,000 gallons of liquid sodium 

(Na) primary coolant, had a closed-loop Na secondary coolant 

system, and produced steam in a tertiary system. The entire system 

was placed in a large containment building. The maximum power 

level was 62.5 MWth, and the EBR-II could supply 20 MWe of 

electric power to INL facilities.

In addition to demonstrating fuel reprocessing and electricity 

production in a liquid metal fast breeder reactor (LMFBR), EBR-II 

also performed irradiation, fuel development, and transient stability 

experiments.

References

INL 2016a 

McFarlane 2016

While OTIB-0054 models the FFTF, a 

fast reactor, and the Hanford N-Reactor, 

a plutonium production reactor, neither 

appears able to adequately envelope the 

EBR reactor conditions of much lower 

power and only intermittent operations. 

Priority Ranking: High

33. Neutron

Radiography Facility 

(NRAD)

ANL-W/EBR-II Complex

1977–

present

NRAD is a 250-kWth, uranium-fueled, light water moderated and 

cooled pool-type, steady-state General Atomics TRIGA reactor, 

located beneath the main hot cell of the Hot Fuel Examination 

Facility (HFEF). It is used primarily for non-destructive neutron 

radiography of irradiated and unirradiated fuel and materials. It has 

limited experimental space inside the core, but its one beamline was 

supplemented by a second in 1982; both have a direct line of sight to 

the core.

References

Craft et al. 2015

OTIB-0054 explicitly models a TRIGA 

reactor, of which NRAD is a typical 

example. Hence, it would be expected 

that OTIB-0054 would adequately 

envelope the NRAD reactor. 

Priority Ranking: Low
Effective Date:

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Reactor Name & 

Location a

Operation 

Dates Summary Description b Comments/Priority Ranking 

50. Transient Reactor

Test Facility (TREAT)

ANL-W/EBR-II Complex

1959–1994 The TREAT experiment continued research and development for fast 

breeder reactors by simulating excursion and accident conditions 

leading to fuel damage. It was an air-cooled, graphite-moderated and 

reflected, uranium-fueled, thermal spectrum reactor, operating at a 

steady power level of up to about 120 kWth. It could also produce 

pulses of up to 19 gigawatt thermal (GWth) for experimental 

purposes, relying on a very large negative temperature coefficient of 

reactivity to then shut it down safely.

The fuel consisted of a 93% enriched graphite-uranium dioxide 

(UO2) matrix, with zirconium clad fuel assemblies. The core was 

approximately 6-feet diameter x 4-feet high, and contained a 19 x 19 

array of 4 in x 4 in fuel and reflector assemblies.

Several different sodium loops were introduced at different times into 

a portion of the core to perform fuel melt experiments from a 

simulated loss-of-coolant flow accident and transient overpower 

accidents. Over the years, the facility tested LMFBR-type oxide fuel 

elements in support of the Fast Flux Test Facility (FFTF) and the 

Clinch River Breeder Reactor Plant safety analysis and licensing (the 

Clinch River plant was cancelled before it was built). Authorized 

personnel were permitted free access to the Reactor Building during 

shutdown, critical operation, or steady-state operation.

In addition to its fast breeder reactor support mission, TREAT was 

also used for light water and space reactor fuel testing and 

nondestructive testing through neutron radiography (examining

specimens up to 15 feet long).

In its lifetime, the reactor performed 6,604 reactor startups and 2,884 

transient irradiations.

References

Bumgardner 2014

INL 2016b

TREAT was an experimental reactor, 

testing different fuels in steady state and 

pulsed mode. While the OTIB-0054 

TRIGA reactor also operates in steady 

state and pulsed mode, TREAT had a 

considerably different core and was 

graphite rather than water moderated. 

Further study would have to be done to 

determine if any of the reactors of 

OTIB-0054 would adequately model 

TREAT.

Priority Ranking: Medium
Effective Date:

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Reactor Name & 

Location a

Operation 

Dates Summary Description b Comments/Priority Ranking 

51. Zero Power Physics

Reactor (Earlier name: 

Zero Power Plutonium 

Reactor) (ZPPR) 

ANL-W/EBR-II Complex

1969–1992 ZPPR was a very low-power, air-cooled, split-table (one table 

movable and the other stationary) critical facility that was used to 

mock up other fast neutron spectrum reactors in support of the overall 

breeder reactor program. It was the largest in the series of split-table 

assemblies that were designed to support the plutonium-fueled liquid 

metal fast breeder reactor program by testing the neutronics of 

reactors of many different sizes, compositions, and shapes.

The split-table design facilitated rapid and easy access to the center 

of core and ensured criticality safety when in the open position. Each 

half of the stainless-steel honeycomb structure was 14 ft × 14 ft × 5 

ft. Reactor materials were stacked in drawers, which were then 

inserted into matrix tubes of the reactor structure. Fuel material 

consisted of high enriched uranium and/or plutonium, and the reactor 

was cooled by liquid sodium. The initial fuel loading had a total 

fissile mass (Pu-239, Pu-241, and U-235) of 363 kg (4/18/69). ZPPR 

Assembly 2 had 1,066 kg Pu-239 and Pu-241.

References

Lawroski 1972

The FFTF modeled in OTIB-0054 is a 

sodium-cooled fast reactor, and the 

Hanford N-Reactor bred plutonium. 

However, ZPPR’s core and operating 

conditions were considerably different 

than either of the OTIB reactors. 

ZPPR’s power level was quite low, 

reducing the priority of performing a 

more detailed investigation.

Priority Ranking: Low
Effective Date:

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0 (Draft)

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Reactor Name & 

Location a

Operation 

Dates Summary Description b Comments/Priority Ranking 

52. Zero Power Reactor

No. 3 (ZPR-III)

ANL-W/EBR-I Complex

1955–1970 ZPR-III, which preceded ZPPR, was the smallest of the ZPPR series 

of split-table critical facilities. The split-table design facilitated rapid 

and easy access to the center of core and ensured criticality safety 

when in the open position. It established the feasibility of a sodiumcooled, unmoderated, fast breeder reactor. It was also used to 

determine the accuracy of predicted mass geometries and critical 

measurements for fast reactor core designs, including those for 

Fermi, Rapsodie, and the Southwest Experimental Fast Oxide 

Reactor (SEFOR).

In all, there were a total of 63 different mockups (“assemblies”), with 

some existing for only short time periods and others for hundreds of 

experiments. Initially, ZPR-II used uranium fuel (coated with 

Teflon), but plutonium fuel (coated with stainless steel was 

introduced later, as was mixed oxide fuel. Fuels were either metallic, 

oxide, or carbide compositions. In addition, the facility switched 

from no cooling system to a simple air cooling system after a few 

years.

References

ANL 2016

Davey 1964

The FFTF modeled in OTIB-0054 is a 

sodium-cooled fast reactor, and the 

Hanford N-Reactor bred plutonium. 

However, ZPR-III’s core and operating 

conditions were considerably different 

than either of the OTIB reactors. 

ZPR-III’s power level was quite low, 

reducing the priority of performing a 

more detailed investigation.

Priority Ranking: Low

Notes: 

(a) The list and numbering scheme of the reactors are taken from Stacy 2000.

(b) The primary sources of information for the summary descriptions are: Stacy 2000, ORAUT 2010, and NIOSH 2016. Other significant sources specific to 

particular reactors are listed at the end of each reactor description.